F74 辐射防护监测与评价 标准查询与下载

ANSI/ASTM E1650:2012 乙酸纤维素剂量测定体系的使用操作规程

Practice for Use of Cellulose Acetate Dosimetry Systems

JJG 309-2011 500K～1000K黑体辐射源检定规程

本规程适用于500 K～l 000 K标准和工作黑体辐射源的首次检定、后续检定和使用中检查。500 K～l 000 K黑体辐射源的型式评价中对有关计量性能的要求可参照本规程执行。

Verification Regulation of Blackbody Radiators at the 500 K～1000 K

NF M60-822-2:2011 核能.气态流出物放射性测量.气态流出物和气体排放中氚和C14活性的测定.第2部分:气态流出物或气体排放截留液中氚活性的测定

Nuclear energy - Radioactivity measurement in gaseous effluents - Determination of tritium and C14 activity in gaseous effluents and gas discharge - Part 2 : determination of tritium activity in the trapping solution of gaseous effluents or gas discharge

DIN 25440:2011 按照当地放射剂量率对核设施及设备的控制区内的房间进行分类

Classification of rooms in the controlled area of nuclear facilities and facilities according to local dose rates

ISO 27048:2011 辐射防护.受内部射线照射的员工检测用计量评定

This International Standard specifies the minimum requirements for the evaluation of data from the monitoring of workers occupationally exposed to the risk of internal contamination by radioactive substances. It presents procedures and assumptions for the standardised interpretation of monitoring data, in order to achieve acceptable levels of reliability. Those procedures allow the quantification of exposures for the documentation of compliance with regulations and radiation protection programmes. Limits are set for the applicability of the procedures in respect of the dose levels above which more sophisticated methods will have to be applied. This International Standard addresses the following: a) procedures for dose assessment based on reference levels for routine and special monitoring programmes; b) assumptions for the selection of dose-critical parameter values; c) criteria for determining the significance of monitoring results; d) interpretation of workplace monitoring results; e) uncertainties arising from sampling, measurement techniques and working conditions; f) the special topics of 1) interpretation of multiple data arising from different measurement methods at different times, 2) handling data below the decision threshold, 3) rogue data, and 4) calculation of doses to the embryo/foetus and infant; g) reporting/documentation; h) quality assurance. It is not applicable to the following: - dosimetry for litigation cases; - modelling for the improvement of internal dosimetry; - the potential influence of decorporation measures (e.g. administration of chelating agents); - the investigation of the causes or implications of an exposure; - dosimetry for contaminated wounds.

Radiation protection. Dose assessment for the monitoring of workers for internal radiation exposure

ANSI N43.1-2011 粒子加速器的设计和操作的辐射安全

This Standard applies to all phases of the accelerator facility life cycle including design, installation, commissioning, operation, maintenance, upgrade and decommissioning. This Standard specifies requirements and recommendations for both the radiation safety program management and technical aspects.

Radiation Safety for the Design and Operation of Particle Accelerators

ANSI N13.1-2011 核设施管道中释放的航空放射性物质的取样监测释放

This standard sets forth guidelines and performance-based criteria for the design and use of systems for sampling the releases of airborne radioactive substances from the ducts and stacks of nuclear facilities.

Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities

ANSI N13.41-2011 执行多个剂量测定的标准

The criteria in this standard provide guidance for when to monitor with multiple dosimeters and where to place such dosimeters, and the interpretation and recording of results after the dosimeters are processed or evaluated.

Criteria for Performing Multiple Dosimetry

ANSI N43.14-2011 货物安检用有源探询系统的辐射安全,能量达100 MeV

This proposed standard establishes radiation safety guidelines, policies and procedures for the safe uses of Active Interrogation Systems so that the operators of these systems and members of the general public, who are in the vicinity of these systems, are protected from unnecessary exposure to neutron (and resulting gamma) radiation and bremsstrahlung (high energy photons). The intent is to ensure that the exposures are well within the regulatory limits.

Radiation Safety for Active Interrogation Systems for Security Screening of Cargo, Energies Up to 100 MeV

BS ISO 2889:2010 从核反应堆和核设施管道中对放射性物质进行空气传播取样

Sampling airborne radioactive materials from the stacks and ducts of nuclear facilities

JIS Z 4514:2010 校准剂量组织计和剂量当量计及测定作为β辐射能量和入射角函数的β能量响应

Calibration of absorbed dose to tissue meters and dose equivalent meters and the determination of their response as a function of beta radiation energy and angle of incidence

BS ISO 28218:2010 辐射保护.放射生物测定用性能标准

Radiation protection. Performance criteria for radiobioassay

BS PD CLC/TR 50484:2009 屏蔽室操作建议

This Technical Report applies to shielded enclosures used for EMC testing which are to be validated according to the EN 50147 series of standards and the corresponding international standards. The object of this report is to give guidance to the selection of the shielding materials and components. The frequency range for this document is 10 kHz to 40 GHz.

Recommendations for shielded enclosures

IEEE N42.47-2010 X射线和伽马射线人身保安系统成像性能测量用美国国家标准

IMPORTANT NOTICE: This standard is not intended to ensure safety, security, health, or environmental protection. Implementers of the standard are responsible for determining appropriate safety, security, environmental, and health practices or regulatory requirements. This IEEE document is made available for use subject to important notices and legal disclaimers. These notices and disclaimers appear in all publications containing this document and may be found under the heading “Important Notice” or “Important Notices and Disclaimers Concerning IEEE Documents.” They can also be obtained on request from IEEE or viewed at http://standards.ieee.org/IPR/ disclaimers.html. This standard applies to security screening systems that utilize x-ray or gamma radiation and are used to inspect people who are not inside vehicles, containers, or enclosures. Specifically, this standard applies to systems used to detect objects carried on or within the body of the individual being exposed. The following types of systems are included in the scope of this standard: ⎯ Systems designated as fixed, portal, re-locatable, transportable, mobile, or gantry. ⎯ Systems employing detection of primary radiation (transmission systems), or scatter radiation (backscatter systems), or a combination of both. ⎯ Systems that are primarily imaging but that also may have complementary features such as material discrimination, or automatic active or passive threat alerts. This standard will not address how to test these complementary features.

American National Standard for Measuring the Imaging Performance of X-ray and Gamma-ray Systems for Security Screening of Humans

NF M62-236:2010 核设施的烟道和通风管道中的放射性物质取样.

Sampling airborne radioactive materials from the stacks and ducts of nuclear facilities.

ASTM E2232-10 辐射处理设备的吸收剂量计算用数学方法的选择和使用的标准指南

Use as an Analytical Tool8212;Mathematical methods provide an analytical tool to be employed for many applications related to absorbed dose determinations in radiation processing. Mathematical calculations may not be used as a substitute for routine dosimetry in some applications (for example, medical device sterilization, food irradiation). Dose Calculation8212;Absorbed-dose calculations may be performed for a variety of photon/electron environments and irradiator geometries. Evaluate Process Effectiveness8212;Mathematical models may be used to evaluate the impact of changes in product composition, loading configuration, and irradiator design on dose distribution. Complement or Supplement to Dosimetry8212;Dose calculations may be used to establish a detailed understanding of dose distribution, providing a spatial resolution not obtainable through measurement. Calculations may be used to reduce the number of dosimeters required to characterize a procedure or process (for example, dose mapping). Alternative to Dosimetry8212;Dose calculations may be used when dosimetry is impractical (for example, granular materials, materials with complex geometries, material contained in a package where dosimetry is not practical or possible). Facility Design8212;Dose calculations are often used in the design of a new irradiator and can be used to help optimize dose distribution in an existing facility or radiation process. The use of modeling in irradiator design can be found in references (3-9). Validation8212;The validation of the model should be done through comparison with reliable and traceable dosimetric measurements. The purpose of validation is to demonstrate that the mathematical method makes reliable predictions of dose and other transport quantities. Validation compares predictions or theory to the results of an appropriate experiment. The degree of validation is commensurate with the application. Guidance is given in the documents referenced in Annex A2. Verification8212;Verification is the confirmation of the mathematical correctness of a computer implementation of a mathematical method. This can be done, for example, by comparing numerical results with known analytic solutions or with other computer codes that have been previously verified. Verification should be done to ensure that the simulation is appropriate for the intended application. Refer to 3.1.23.1. Note 28212;Certain applications of the mathematical model deal with Operational Qualification (OQ), Performance Qualification (PQ) and process control in radiation processing such as the sterilization of healthcare products. The application and use of the mathematical model in these applications may have to meet regulatory requirements. Refer to Section 6 for prerequisites for application of a mathematical method and Section 8 for requirements before routine use of the mathematical method. Uncertainty8212;An absorbed dose prediction should be accompanied by an estimate of overall uncertainty, as it is with absorbed-dose determination (refer to ISO/ASTM 51707 and NIST Technical Note 1297). In many cases, absorbed-dose measurement helps to establish the uncertainty in the dose calculation. This guide should not be used as the only reference in the selection and use of mathematical models. The user is encoura......

Standard Guide for Selection and Use of Mathematical Methods for Calculating Absorbed Dose in Radiation Processing Applications

ANSI/ISO/ASTM 51261E 1261:2010 辐射加工用剂量测定系统的选择和校准用指南

GUIDE FOR SELECTION AND CALIBRATION OF DOSIMETRY SYSTEMS FOR RADIATION PROCESSING

ASTM E185-10 轻水中型核动力反应堆罐监测计划设计的标准实施规程

Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel. The methodology to be used in estimation of neutron exposure obtained for reactor vessel surveillance programs is defined in Guide E482. The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materials in addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure conditions and material characteristics will determine their applicability for predicting radiation effects.1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life, but the procedure described may provide guidance for developing such a surveillance program. 1.5 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. Note 18212;The increased complexity of the requirements for a light-water moderated nuclear power reactor vessel surveillance program has necessitated the separation of the requirements into three related standards. Practice E185 describes the minimum requirements for a surveillance program. Practice E2215 describes the procedures for testing and evaluation of surveillance capsules removed from a surveillance program as defined in the current or previous editions of Practice E185. Guide E......

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

ASTM C1726/C1726M-10 无源γ射线测量用模型的标准使用指南

The following methods assist in demonstrating regulatory compliance in such areas as safeguards (Special Nuclear Material), inventory control, criticality control, decontamination and decommissioning, waste disposal, holdup and shipping. This guide can apply to the assay of radionuclides in containers, whose gamma-ray absorption properties can be measured or estimated, for which representative certified standards are not available. It can be applied to in situ measurements, measurement stations, or to laboratory measurements. Some of the modeling techniques described in the guide are suitable for the measurement of fall-out or natural radioactivity homogenously distributed in soil. Source-based efficiency calibrations for laboratory geometries may suffer from inaccuracies due to gamma rays being detected in true coincidence. Modeling can be an advantage since it is unaffected by true coincidence summing effects.1.1 This guide addresses the use of models with passive gamma-ray measurement systems. Mathematical models based on physical principles can be used to assist in calibration of gamma-ray measurement systems and in analysis of measurement data. Some nondestructive assay (NDA) measurement programs involve the assay of a wide variety of item geometries and matrix combinations for which the development of physical standards are not practical. In these situations, modeling may provide a cost-effective means of meeting user’s data quality objectives. 1.2 A scientific knowledge of radiation sources and detectors, calibration procedures, geometry and error analysis is needed for users of this standard. This guide assumes that the user has, at a minimum, a basic understanding of these principles and good NDA practices (see Guide C1592), as defined for an NDA professional in Guide C1490. The user of this standard must have at least a basic understanding of the software used for modeling. Instructions or further training on the use of such software is beyond the scope of this standard. 1.3 The focus of this guide is the use of response models for high-purity germanium (HPGe) detector systems for the passive gamma-ray assay of items. Many of the models described in this guide may also be applied to the use of detectors with different resolutions, such as sodium iodide or lanthanum halide. In such cases, an NDA professional should determine the applicability of sections of this guide to the specific application. 1.4 Techniques discussed in this guide are applicable to modeling a variety of radioactive material including contaminated fields, walls, containers and process equipment. 1.5 This guide does not purport to discuss modeling for “infinite plane” in situ measurements. This discussion is best covered in ANSI N42.28. 1.6 This guide does not purport to address the physical concerns of how to make or set up equipment for in situ measurements but only how to select the model for which the in situ measurement data is analyzed. 1.7 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the standard. 1.8 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are n......

Standard Guide for Use of Modeling for Passive Gamma Measurements

ASTM E261-10 放射活化法测定中子积分通量、通量率以及频谱的标准实施规程

Transmutation Processes8212;The effect on materials of bombardment by neutrons depends on the energy of the neutrons; therefore, it is important that the energy distribution of the neutron fluence, as well as the total fluence, be determined.1.1 This practice describes procedures for the determination of neutron fluence rate, fluence, and energy spectra from the radioactivity that is induced in a detector specimen. 1.2 The practice is directed toward the determination of these quantities in connection with radiation effects on materials. 1.3 For application of these techniques to reactor vessel surveillance, see also Test Methods E1005. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. Note 18212;Detailed methods for individual detectors are given in the following ASTM test methods: E262, E263, E264, E265, E266, E343, E393, E481, E523, E526, E704, E705, and E854.

Standard Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques